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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 433 records · Page 24
All sourcesNASA NTRSDOE OSTI
DOE OSTI2014

Synthesis and characterization of [BMIM]bromide using microwave-assisted organic synthesis method and its application for dissolution of palm empty fruit bunch

The decrease of cellulose crystallinity index of palm empty fruit bunch is crucial for the next application of cellulose as raw material for various biofuel and its derivatives. The aim of this research is to decrease the cellulose crystallinity index of palm empty fruit bunch using 1-butyl-3-methylimidazoliumbromide or [BMIM] bromide which has been synthesized using Microwave-Assisted Organic Synthesis (MAOS) method. Conventional reaction method has also been carried out to synthesize [BMIM]bromide for comparison as well. The characterization of synthesized product using FTIR, {sup 1}H-NMR, {sup 13}C-NMR and LC-MS showed that these reactions have been carried out successfully. The results showed that MAOS method is up to 90% faster in producing [BMIM]bromide compare to the conventional method. The application of [BMIM]bromide for dissolution of palm empty fruit bunch showed that cellulose and lignin could be extracted using stirring process for 20 hours. The decrease of cellulose crystallinity index and its morphology changes were identified using FTIR and Scanning Electron Microscope.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2014

Field-assisted patterned dissolution of silver nanoparticles in phosphate glass

Phosphate glass samples doped with silver ions through a Na⁺-Ag⁺ ion-exchange process were treated in a hydrogen atmosphere at temperatures near 430°C for durations ranging from 4 to 5 h. Such treatment causes metallic silver precipitation at the surface as well as nanoclustering of silver atoms under the surface under conditions very similar to those used for silicate glasses. The presence of silver clusters resulted in a characteristic coloring of the glass and was verified by the observation of a plasmon resonance peak near 410–420 nm in the absorption spectra. Applying a DC voltage between 1.4 and 2 kV at temperatures between 120 and 130°C led to dissolution of the clusters in the area under the positive electrode, thereby bleaching the glass color. The use of a patterned doped-silicon electrode further led to the formation of a 300 nm thick surface relief on the glass surface and of a volume complex permittivity grating extending at least 4 μm under the surface. Such volume complex refractive index gratings may find applications in passive or active (laser) photonic devices in rare-earth doped phosphate glasses, where conventional bulk grating formation techniques have limited applicability.

77 NANOSCIENCE AND NANOTECHNOLOGY↗
DOE OSTI2017

Characterization of undissolved solids from the dissolution of North Anna reactor fuel

Here, samples of undissolved solids (UDS) from the dissolution of North Anna reactor fuel were characterized to investigate the effects of using air or oxygen as the oxidant during tritium removal. The UDS composition data also support the development of a waste form for disposal. There was no discernible effect of the oxidant used during the tritium removal process or the size fraction on the UDS composition. Scanning electron microscopy (SEM) and energy dispersive (x-ray) spectroscopy were used to estimate the oxygen content of the UDS and it was found to be potentially significant, on the order of 30% by mass and 80% by atom.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Dissolution flowsheet for high flux isotope reactor fuel at SRS

H-Canyon at Savannah River Site has started dissolving High Flux Isotope Reactor (HFIR) fuel as part of the Spent Nuclear Fuel (SNF) processing campaign. Each HFIR fuel core contains inner and outer fuel elements fabricated from uranium oxide (U 3 O 8 ) dispersed in a continuous Al phase. Fuels fabricated in this manner, like other SNF’s processed in H-Canyon, dissolve by the same general mechanisms with similar gas generation rates including the production of H 2 . The objective of this work was to identify flowsheet conditions through literature review and laboratory experimentation to safely and efficiently dissolve the HFIR fuel in H-Canyon. In conclusion, the work showed that a full HFIR core could be dissolved using nominally 0.002 M Hg to catalyze the dissolution and the margin between the predicted H 2 concentration and the calculated Lower Flammability Limit (LFL) was greater when the dissolving solution was allowed to boil for 45 min prior to initiating the Hg addition.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Dissolution of used nuclear fuel using a tributyl phosphate/n-paraffin solvent

The complete recycle of used nuclear fuel (UNF) is necessary to recover the energy value from the remaining U and higher actinides; yet, construction of a large-scale facility is extremely expensive. To address this issue, a hybrid process was evaluated in which UNF pretreated for tritium removal is dissolved in a tributyl phosphate (TBP) solvent. The distribution of the actinides and selected fission products was measured between the solvent and residual solids. Dissolution of these elements in the surrogate fuels was very effective. Most fission products from fuel irradiation except for the lanthanide elements would remain in the undissolved solids.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Growth and dissolution of spherical density enhancements in SCDEW cosmologies

Strongly Coupled Dark Energy plus Warm dark matter (SCDEW) cosmologies are based on the finding of a conformally invariant (CI) attractor solution during the early radiative expansion, requiring then the stationary presence of ∼ 1 % of coupled-DM and DE, since inflationary reheating. In these models, coupled-DM fluctuations, even in the early radiative expansion, grow up to non-linearity, as shown in a previous associated paper. Such early non-linear stages are modelized here through the evolution of a top-hat density enhancement. As expected, its radius R increases up to a maximum and then starts to decrease. Virial balance is reached when the coupled-DM density contrast is just 25–26 and DM density enhancement is O(10 %) of total density. Moreover, we find that this is not an equilibrium configuration as, afterwards, coupling causes DM particle velocities to increase, so that the fluctuation gradually dissolves. We estimate the duration of the whole process, from horizon crossing to dissolution, and find z {sub horizon}/ z {sub erasing} ∼ 3 × 10{sup 4}. Therefore, only fluctuations entering the horizon at z ∼< 10{sup 9}–10{sup 10} are able to accrete WDM with mass ∼ 100 eV—as soon as it becomes non-relativistic—so avoiding full disruption. Accordingly, SCDEW cosmologies, whose WDM has mass ∼ 100 eV, can preserve primeval fluctuations down to stellar mass scale.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2016

Asymmetric collapse by dissolution or melting in a uniform flow

An advection-diffusion-limited dissolution model of an object being eroded by a two-dimensional potential flow is presented. By taking advantage of the conformal invariance of the model, a numerical method is introduced that tracks the evolution of the object boundary in terms of a time-dependent Laurent series. Simulations of a variety of dissolving objects are shown, which shrink and collapse to a single point in finite time. The simulations reveal a surprising exact relationship, whereby the collapse point is the root of a non-Analytic function given in terms of the flow velocity and the Laurent series coefficients describing the initial shape. This result is subsequently derived using residue calculus. The structure of the non-Analytic function is examined for three different test cases, and a practical approach to determine the collapse point using a generalized Newton-Raphson root-finding algorithm is outlined. These examples also illustrate the possibility that the model breaks down in finite time prior to complete collapse, due to a topological singularity, as the dissolving boundary overlaps itself rather than breaking up into multiple domains (analogous to droplet pinch-off in fluid mechanics). In conclusion, the model raises fundamental mathematical questions about broken symmetries in finite-Time singularities of both continuous and stochastic dynamical systems.

42 ENGINEERING↗
DOE OSTI2020

Monitoring the Caustic Dissolution of Aluminum Alloy in a Radiochemical Hot Cell Using Raman Spectroscopy

Chemical processing of highly radioactive materials commonly takes place in heavily shielded hot cells. The remote, real-time monitoring of chemical processing streams via optical spectroscopic techniques in hot cells may be particularly useful. Here in this paper, we describe the implementation of Raman spectroscopy and chemometric analysis to monitor the dissolution of aluminum-clad targets containing irradiated aluminum–neptunium oxide cermet pellets in caustic solutions in a hot cell environment. Partial least squares regression analysis was used to generate calibration models to quantify the concentration of dissolved aluminum, nitrate, and hydroxide in solutions within the radiochemical hot cell. This work explored a systematic approach to optimize a matrix of calibration standards using a D-optimal experimental design. The Design of Experiments-based regression model, in comparison to more traditional analytical approaches, was found to be the more practical method for building calibration models, with fewer samples, to obtain informative analytical data from Raman spectra.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Silicate coprecipitation reduces green rust crystal size and limits dissolution-precipitation during air oxidation

Green rusts (GR) are mixed-valence iron (Fe) hydroxides which form in reducing redox environments like riparian and wetland soils and shallow groundwater. In these environments, silicon (Si) can influence Fe oxides’ chemical and physical properties but its role in GR formation and subsequent oxidative transformation have not been studied starting at initial nucleation. Green rust sulfate [GR(SO 4 )] and green rust carbonate [GR(CO 3 )] were both coprecipitated from salts by base titration in increasing % mol Si (0, 1, 10, and 50). The minerals were characterized before and after rapid (24 h) aqueous air-oxidation by x-ray diffraction (XRD), scanning electron microscopy (SEM), Fe extended x-ray absorption fine structure spectroscopy (EXAFS), and N 2 -BET surface area. Results showed that only GR(SO4) or GR(CO3) was formed at every tested Si concentration. Increasing % mol Si caused decreased plate size and increased surface area in GR(CO3) but not GR(SO4). GR plate basal thickness was not changed at any condition indicating a lack of Si interlayering. Air oxidation of GR(SO4) at all % mol Si contents transformed by dissolution and reprecipitation into lepidocrocite and goethite, favoring ferrihydrite with higher % Si content. Air oxidation of GR(CO3) transformed into magnetite and goethite but increasing Si caused GR to oxidize while retaining its hexagonal plate structure via solid-state oxidation. Our results indicate that Si has the potential to cause GR to form in smaller particles and upon air oxidation, Si can either stabilize the plate structure or alter transformation to ferrihydrite.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Chlorite, Biotite, Illite, Muscovite and Feldspar Dissolution Kinetics at Variable pH and Temperatures up to 280 deg C

Chemical reactions pose an important but poorly understood threat to EGS long-term success because of their impact on fracture permeability. This report summarizes the dissolution rate equations for layered silicates where data were lacking for geothermal systems. Here we report updated rate laws for chlorite (Carroll and Smith 2013), biotite (Carroll and Smith, 2015), illite (Carroll and Smith, 2014), and for muscovite. Also included is a spreadsheet with rate data and rate equations for use in reactive transport simulators.

15 GEOTHERMAL ENERGY↗
DOE OSTI2020

Dissolution of poorly soluble uranyl phosphate phases in the Metaautunite Subgroup under uranyl peroxide cage cluster forming conditions

Uranyl phosphate minerals are widespread in uranium deposits and usually exhibit very low solubility in aqueous systems. Uranyl phosphates of the autunite group and metaautunite subgroup impact the mobility of uranium in the environment and have inspired groundwater remediation strategies that emphasize their low solubility. The importance of soluble uranium-bearing macro-anions, including nanoscale uranyl peroxide cage clusters, is largely unexplored relative to solubilization of normally low-solubility uranium minerals. Eight synthetic analogs of metaautunite subgroup minerals have been prepared and placed in various alkaline aqueous solutions containing hydrogen peroxide and tetraethylammonium hydroxide. Each uranyl phosphate studied has a topologically identical anionic sheet of uranyl square bipyramids and phosphate tetrahedra combined with various cations (Li + , Na + , K + , Rb + , Cs + , Mg 2+ , Ca 2+ , Ba 2+ ) and water in the interlayer. Uranyl peroxides formed under many of the experimental conditions examined, including solid studtite [(UO 2 )(O 2 )(H 2 O) 2 ](H 2 O) 2 and soluble uranyl peroxide cage clusters containing as many as 28 uranyl ions. Uranyl phosphate solids in contact with solutions in which uranyl peroxide cage clusters formed dissolved extensively or completely. The greatest dissolution of uranyl phosphates occurred in systems that contained cations with larger hydrated radii, Li + and Na + . The details of the uranium speciation in solution depended on the pH and counter cations provided from the interlayers of the uranyl phosphate solids.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2011

Radioactive Waste Management: Study of Spent Fuel Dissolution Rates in Geological Storage Using Dosimetry Modeling and Experimental Verification

This research will provide improved predictions into the mechanisms and effects of radiolysis on spent nuclear fuel dissolution in a geological respository through accurate dosimetry modeling of the dose to water, mechanistic chemistry modeling of the resulting radiolytic reactions and confirmatory experimental measurements. This work will combine effort by the Nuclear Science and Engineering Institute (NSEI) and the Missouri University Research Reactor (MURR) at the University of Missouri-Columbia, and the expertise and facilities at the Pacific Northwest National Laboratory (PNNL).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Sample Dissolution and Purification for 230 Th- 234 U Age-Dating

This is the standard operating procedure used by the Isotope Ratio Mass Spectrometry Group of the Chemical Sciences Division at LLNL for the preparation of a sample of uranium oxide or uranium metal for 230 Th- 234 U age-dating. The method described here includes the dissolution of a sample of uranium oxide or uranium metal, preparation of a secondary dilution, spiking of separate aliquots for uranium and thorium isotope dilution measurements, and purification of uranium and thorium aliquots for mass spectrometry. This SOP may be applied to uranium samples of unknown purity as in a nuclear forensic investigation, and also to well-characterized samples such as, for example, U 3 O 8 and U-metal certified reference materials. The sample of uranium is transferred to a quartz or PFA vial, concentrated nitric acid is added and the sample is heated on a hotplate at approximately 100°C for several hours until it dissolves. The sample solution is diluted with water to make the solution approximately 4 M HNO 3 and hydrofluoric acid is added to make it 0.05 M HF. A secondary dilution of the primary uranium solution is prepared. Separate aliquots for uranium and thorium isotope dilution measurements are taken and spiked with 233 U and 229 Th, respectively. The spiked aliquot for uranium isotope dilution analysis is purified using EiChrom UTEVA resin. The spiked aliquot for thorium isotope dilution analysis is purified by, first, a 1.8 mL AG1x8 resin bed in 9 M HCl on which U adsorbs and Th passes through; second, adsorbing Th on a 1 mL AG1x8 resin bed in 8 M HNO 3 and then eluting it with 9 M HCl followed by 0.1 M HCl + 0.005 M HF; and third, by passing the Th through a final 1.0 mL AG1x8 resin bed in 9 M HCl. The mass spectrometry is performed using the procedure 'Th and U Mass Spectrometry for 230 Th- 234 U Age Dating'.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2013

Rapid Automated Dissolution and Analysis Techniques for Radionuclides in Recycle Process Streams

The analysis of process samples for radionuclide content is an important part of current procedures for material balance and accountancy in the different process streams of a recycling plant. The destructive sample analysis techniques currently available necessitate a significant amount of time. It is therefore desirable to develop new sample analysis procedures that allow for a quick turnaround time and increased sample throughput with a minimum of deviation between samples. In particular, new capabilities for rapid sample dissolution and radiochemical separation are required. Most of the radioanalytical techniques currently employed for sample analysis are based on manual laboratory procedures. Such procedures are time- and labor-intensive, and not well suited for situations in which a rapid sample analysis is required and/or large number of samples need to be analyzed. To address this issue we are currently investigating radiochemical separation methods based on extraction chromatography that have been specifically optimized for the analysis of process stream samples. The influence of potential interferences present in the process samples as well as mass loading, flow rate and resin performance is being studied. In addition, the potential to automate these procedures utilizing a robotic platform is evaluated. Initial studies have been carried out using the commercially available DGA resin. This resin shows an affinity for Am, Pu, U, and Th and is also exhibiting signs of a possible synergistic effects in the presence of iron.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Analysis of maximum allowable fragment heights during dissolution of high flux isotope reactor fuel in an h-canyon dissolver

As part of the Spent Nuclear Fuel (SNF) processing campaign, H-Canyon is planning to begin dissolving High Flux Isotope Reactor (HFIR) fuel in late FY17 or early FY18. Each HFIR fuel core contains inner and outer fuel elements which were fabricated from uranium oxide (U 3 O 8 ) dispersed in a continuous Al phase using traditional powder metallurgy techniques. Fuels fabricated in this manner, like other SNF’s processed in H-Canyon, dissolve by the same general mechanisms with similar gas generation rates and the production of H 2 . The HFIR fuel cores will be dissolved using a flowsheet developed by the Savannah River National Laboratory (SRNL) in either the 6.4D or 6.1D dissolver using a unique insert. Multiple cores will be charged to the same dissolver solution maximizing the concentration of dissolved Al. The recovered U will be down-blended into low-enriched U for subsequent use as commercial reactor fuel. During the development of the HFIR fuel dissolution flowsheet, the cycle time for the initial core was estimated at 28 to 40 h. Once the cycle is complete, H-Canyon personnel will open the dissolver and probe the HFIR insert wells to determine the height of any fuel fragments which did not dissolve. Before the next core can be charged to the dissolver, an analysis of the potential for H 2 gas generation must show that the combined surface area of the fuel fragments and the subsequent core will not generate H 2 concentrations in the dissolver offgas which exceeds 60% of the lower flammability limit (LFL) of H 2 at 200 °C. The objective of this study is to identify the maximum fuel fragment height as a function of the Al concentration in the dissolving solution which will provide criteria for charging successive HFIR cores to an H-Canyon dissolver.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Uranium Zirconium Alloy Dissolution

This Special Instructions provides a standard method to dissolve irradiated or unirradiated UZr alloys. The scope of this procedure was developed to facilitate total dissolution of UZr solid samples with and without cladding materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Bench-Scale Electrolytic Dissolution of Quarter-Scale FCA Cans

In 2016, the Savannah River National Laboratory (SRNL) led, in support of and under sponsorship of the Department of Energy’s National Nuclear Security Administration (DOE/NNSA) Office of Material Management and Minimization (M3), the removal and transfer of the plutonium based Fast Critical Assembly (FCA) fuel from the Japan Atomic Energy Agency (JAEA) Tokai facility to the Savannah River Site (SRS). The team also included JAEA, multiple organizations in Savannah River Nuclear Solutions (SRNS), International Nuclear Services, and many other entities. The FCA fuel removal project completion was a key deliverable for M3 to the 2016 Nuclear Security Summit and constituted the largest inventory of weapons-usable plutonium removed under the nonproliferation program. The FCA materials consist of thousands of stainless steel (SS) clad plates and hundreds of SS clad rods. The FCA fuel elements were packaged in a carrier can and stored at SRS pending disposition of the fuel. Following an assessment of candidate disposition options, SRNS identified electrolytic dissolution (ED) as the most promising disposition option for the FCA plates and their preferred option was endorsed by DOE. This option entails electrochemically dissolving the entire FCA carrier can with fuel elements and was based on bench-scale laboratory testing and historical work on processing SS-clad and zirconium-clad uranium-based fuel in the H-Canyon electrolytic dissolver (last operated in 1980). The FCA plate consists of a plutonium-aluminum metal alloy core hermitically sealed in SS cladding.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗